Creep properties of oxide and carbide fuels under irradiation

The creep behaviour of uranium dioxide and uranium carbide has been examined in both bend and compression experiments in DIDO Materials Test Reactor. In UO₂ no significant variation in creep rate with dose and temperature occured above ~10²⁵ fissions m^?3 between 450°C and 1230°C, the high strain rates measured in compression at low doses being largely attributable to pore sintering. Both a linear rating and stress dependence were observed up to 40 MNm^?2 and creep rates were found to be independent of grain size. At higher doses ($\text{>6 × 10}{}^{26}\text{fissions m}{}^{\mathrm{?3}}$) transient strains were incurred on varying stress and temperature due to the development of grain boundary gas bubbles. This also resulted in a six fold increase in the radiation creep constant between $\text{6 × 10}{}^{26}$ and $\text{1.2 × 10}{}^{27}$ fissions m^?3. A similar pattern of behaviour with respect to rating and stress was observed in hyperstoichiometric UC between 450 and 800°C up to $\text{1 × 10}{}^{27}$ fissions m^?3. However the nominally steady state creep rate was a factor 8 lower than in UO₂ irradiated under the same conditions. The experimental results also suggest that the primary creep contribution to the initial strain in compression is much higher than in UO₂. There was no evidence of either transient strain on changing stress or of an increasing creep rate at high doses. The experimental observations are reported and discussed in relation to models for irradiation induced low temperature creep in ceramic fuels.     

Electrical conductivity and thermoelectric power in irradiated UO2 + x

Two kinds of UO_{2 + x}, the O/U ratios of which were 2.002 and 2.004, respectively, were irradiated to a dose range between 1.14 × 10¹⁴ and 1.90 × 10¹⁸ fissions/cm³, and electrical conductivity changes were measured. A steep decrease in conductivity was observed with increasing dose up to 1 × 10¹⁵ fissions/cm³, a gradual increase followed between 1 × 10¹⁵ and 1 × 10¹⁸ fissions/cm³ and above this dose the conductivity abruptly increased. Thermoelectric power measurements were also carried out for the specimens irradiated in the dose range up to 1.90 × 10¹⁸ fissions/cm³. It might be suggested that p-type conduction contributes to the electrical conductivity in irradiated specimens up to 1.90 × 10¹⁸ fissions/cm³.     

Observations on the low-temperature creep properties of uranium carbide under irradiation

The creep behaviour of 97% dense hyperstoichiometric UC has been examined during irradiation in three-point bend tests carried out at 450°C up to a dose of $\text{1.65 × 10}{}^{26}\text{fissions/m}{}^3$. A rapid decrease in measured strain rate with dose was observed at each stress level, nominally steady-state creep being established above $\text{≈ 1 × 10}{}^{26}\text{fissions/m}{}^3$ when the creep rate was a factor of 8 lower than that observed in UO₂ irradiated under identical conditions. Creep rates were found to be directly proportional to stress at high doses. Comparison of results from this test with data from other experiments up to $\text{2 × 10}{}^{25}\text{fissions/m}{}^3$ in compression and tension indicates little variation in the radiation-creep constant between 450°C and 800°C. The creep rate for UC, much lower than that observed in UO₂, is consistent with recently reported determinations of the effective uranium self-diffusion coefficients under irradiation in those materials.     

Release of xenon from low-dose irradiated thoria pellets

The kinetics of post-irradiation thermal release of Xe-133 from ThO₂-0.1% UO₂ pellets of different densities, ranging from 67 to 93% theoretical density have been studied. The initial burst release (?₀) and apparent diffusivity ($\text{D'}$) decrease considerably in the case of high-density pellets (above 85% TD), which showed an abnormal increase in the closed porosity and sharp decrease in the open porosity, resulting in a large decrease in surface area. Activation energies for the release of Xe-133 have been evaluated and the possible mechanisms of release are discussed. The effect of Y₂O₃ doping (0.25 mole%) on xenon release has been studied to improve the understanding of the mechanism of rare gas migration in thoria lattice. The small decrease of diffusivity (~2–4 times) in doped pellets shows the absence of migration through cation or anion vacancies. The possible mechanism of release is discussed.     

Effect of Fast Neutron Irradiation on the Properties of Boron Carbide Pellet

Boron carbide pellets were irradiated in the experimental fast reactor “JOYO” to ¹⁰B burnup of up to 170x10²⁶cap/m³, fluences of 2x10²⁶/m²(E>0.1MeV), and maximum temperatures of about 1,200°C. Post irradiation examinations were made of microstructural changes, helium release, swelling, and thermal conductivity.

Boron carbide pellets irradiated to high burnups developed extensive cracking. Helium release from the pellets was initially low, but enhanced helium release was observed at high burnups and high temperatures. The swelling linearly increased with burnup, and when boron carbide was irradiated at high temperatures, the swelling rate began to decrease corresponding to the beginning of enhanced helium release. The correlation between swelling and the helium release was studied and the swelling was interpreted in terms of accumulation of helium in the boron carbide pellet. The thermal conductivity of the boron carbide pellets decreased rapidly by neutron irradiation accompanied with loss of temperature dependence.     

Some observations of density and porosity changes in UO2 fuel irradiated in water-cooled reactors

A technique has been developed for the hot-cell measurement of the apparent density of irradiated UO₂ fuel after extraction from a fuel pin. A single determination is accurate to $\text{± 3 \%}$ at the 95 % confidence limit. The method has been applied to fuel irradiated in thermal neutron fluxes in the Winfrith SGHWR and in the Halden BWR. Material has been examined at ratings of 1–51 W/g and in the burn-up range $\text{0.09–5.79 × 10}{}^{20}\text{fissions/cm}$. It is concluded that pellets with peak temperatures below 1100°C densify during irradiation, but at higher temperatures the pellets begin to swell. Fuel micrography has shown that the densification is principally due to the loss of micropores with a temperature dependency given by an activation energy of 5200 cal/mol. Above 1000°C the densification is masked by the formation and growth of intergranular fission gas bubbles, whose volume may exceed that of the manufactured pores which have sintered. In solid fuel pellets central swelling did not balance densification in the cooler rim until the fuel centre temperature exceeded 1700°C.     

Re-solution-controlled fission-gas behavior in UO2 irradiated in a fast flux

Uranium dioxide irradiated in a fast neutron flux to a burnup of $\text{2 × 10}{}^{20}$ fissions/cm³ between 650 and 1400°C has been examined by transmission- and scanning-electron microscopy and replication metallography. The fission-gas distribution in the fuel matrix and grain boundaries has been characterized as a function of irradiation temperature and fission rate. The majority of fission gas produced even at the highest irradiation temperature was in the UO₂ matrix either in solution or in the form of bubbles < 20 Å in diameter. The results are explained on the basis of an irradiation-induced re-solution mechanism whereby fission gas from within bubbles is reinjected into metastable solution in the UO₂ lattice. Calculated fission-gas solubilities are given as a function of temperature for 10¹³, $\text{3 × 10}{}^{13}$, and 10¹⁴ fissions/cm³ · sec, and, based on these results, it is concluded that the re-solution process is operative over a substantial fuel volume of both light-water-reactor and fast-breeder-reactor oxide fuels.     

Densification behaviour of ThO2–PuO2 pellets with varying PuO2 content using dilatometry

The shrinkage behaviour of ThO₂, ThO₂–30%PuO₂, ThO₂–50%PuO₂ and ThO₂–75%PuO₂ pellets has been studied using a dilatometer in inert (Ar) and reducing atmospheres (Ar–8%H₂). The effects of dopants of CaO and Nb₂O₅ on shrinkage of the oxides of the above Pu/(Pu+Th) ratios were also studied. Out of the two dopants studied, CaO was found to give larger shrinkage for all the Pu/(Pu+Th) ratios covered in this study. It was also found that the shrinkage was marginally larger in Ar–8%H₂ than in Ar atmosphere. Addition of PuO₂ to ThO₂ enhanced sintering. This was found to be true for both the dopants. During the sintering of ThO₂, a prominent peak was observed in the shrinkage curve at around 100–300 °C. This peak was attributed to the pressure increase of the trapped gases which subsequently release at high temperatures.     

Effects of Temperature Cycling and Heating Rate on Fission Gas Release of BWR Fuels

The effects of temperature cycling and heating rate on the release behavior of ⁸⁵Kr have been studied for U0₂ pellets irradiated in a commercial BWR during 3 and 4 cycles (burn-up: 23 and 28GWd/t), by using a post irradiation annealing technique. In addition, characteristics of intergranular bubbles in base-irradiated and annealed specimens (burn-up: 6~28GWd/t) have been examined by SEM fractography.

No significant difference in the release of ⁸⁵Kr was observed between the cyclic heating from 700 to 1,400°C and isothermal heating at 1,400°C. The maximum release rate of ⁸⁵Kr during heating up to 1,800°C became lower with decreasing heating rate in the range of 0.03–10°C/s, while its cumulative fractional releases were about 20~30%, almost independent of heating rate. The fractional coverage of the grain face area occupied by intergranular bubbles saturated around 40~50 for the specimens annealed at 1,600-1,800°C, independent of specimen burn-up and heating conditions (temperature, heating rate and duration). A relationship between intergranular bubble concentration N_g per unit area of grain face and average bubble diameter d_g was expressed as N_g∝d_{g 2.1}     

Fission Gas Release from Nuclear Fuels, (II)

The release behavior of fission gases in U-metal, UO₂ and uranium carbides, irradiated at a relatively low temperature (below 100°C) to low dosage, was studied by out-of-pile experiments.

It was found that fission gas (¹³³Xe) released from a specimen by fission fragment recoil is mostly captured in the wall of the irradiating capsule or in the capsule support material.

The amount of fission gas released into the void space of the capsule is proportional to the surface area and to the fuel burn-up, and is controlled by a knock-out release mechanism. The number of U atoms considered to take part in the knock-out mechanism by evaporation or displacement due to the intrusion of a recoil fission fragment, is estimated to be 1.4×10⁵~2.7×10⁵ atoms for U-metal and 5×10⁴~10×10⁴ atoms for UO₂ and uranium carbides.     

The effect of additives on the irradiation behaviour of UO2

Samples of UO₂ doped with small amounts of Nb₂O₅ or La₂O₃, and having various grain sizes, have been irradiated at 1500°C to 0.1% FIMA. At this low burn-up, gas release and swelling measurements show no dependence on dopant, but the Booth model prediction of swelling proportional to reciprocal grain size has been verified. Gas release does not fit the simple Booth model at the low releases measured, and shows a dependence on sample density, and hence surface area only. A model has been derived to explain these results. The rare gas diffusion coefficient in UO₂ at 1500°C has been measured to be $\text{1.6 × 10}{}^{\mathrm{?19}}$ m²/s.     

Method for Determining Degassing Constant of Xenon in Liquid Sodium

A method for determining the xenon degassing constant from liquid sodium to the cover gas by measuring the ¹³⁵Xe and ^135mXe release rate ratios, which were produced by the decay of ¹³⁵I in liquid sodium, was investigated using the Toshiba inpile fission product loop. Release rate ratios of ¹³⁵Xe to ^135mXe showed values of 9.9~7.6 at a sodium temperature range of 240~390°C and degassing constant of 1.2 × 10^?3-2.6 × 10^?3 sec^?1, which correspond to degassing half-life of 9.7~4.4 min, were obtained. This method is found to be effective for the determination of fission gas degassing constant, which has a corresponding half-life between several minutes to 40 min and will be useful for the study of fission gas transport phenomena in liquid sodium systems.     

Preparation of Uranium-233

Thorium dioxide amounting to 1.3 kg was irradiated for 241 days at an average neutron flux of 6×10¹⁰n/cm²'sec, and allowed to stand for 1.2yr. This cooling lowered the γ-activity of the irradiated ThO₂ sufficiently to permit handling without special shielding. Nine milligrams of ²³³U were extracted with TBP in dodecane, and determined by α-ray counting. The mass ratio of the uranium isotopes obtained was determined by mass spectrometry. It was indicated that the ²³³U was contaminated with natural uranium.     

Sol-gel extrusion process for fabrication of (Th,U)O2 recycle fuel

An extrusion process based on sol-gel derived paste has been developed for the production of thoria recycle fuel as an alternative to the conventional powder-compaction/sintered-pellet route. Crack-free, high-density (9.7 Mg/m³) extruded slugs have been fabricated from sol-gel pastes prepared from ThO₂ powder (denitrated at 600°C) having a moisture content of ~ 16%, and mixed with phenolic resin. The extruded slugs were finally sintered at 1600°C. The effects of thorium-nitrate denitration temperature and organic binder addition on the quality of the fuel slugs are discussed in the paper.     

The mechanisms of fission gas release from (Th,U)O2

The releases of xenon from three (Th, U)O₂ specimens with different U contents were measured over a wide range of fission dose from $\text{2.9 × 10}{}^{19}$ to $\text{2.2 × 10}{}^{22}$ fissions m^?3 by using a post-irradiation technique. The releases were found to decrease with dose and to level off at higher doses. Measurements of the changes in lattice parameter and specific surface area of the same specimens enabled one to conclude that the decrease in release originates in the trapping of xenon by the vacancies and vacancy clusters induced by fission fragments. And the release mechanisms of fission gas were proposed based on the proper evaluation of the observation on radiation damage and recovery in oxide fuel.     

Low-energy helium implantation of aluminum

A series of 20 keV He⁺ implantations was conducted on well-annealed MARZ grade aluminum at fluxes of 6 × 10¹⁴ and 8 × 10¹³He⁺/cm² sec. Three distinct, temperature dependent He release mechanisms were found by He re-emission measurements during implantation, and by subsequent SEM and TEM investigations. At 0.08 of the melting temperature (T_m) gas re-emission rose smoothly after a critical dose of 3 × 10¹⁷He⁺/cm², with extensive blistering. The intermediate temperature range (~0.3 T_m) was characterized by repeated flake exfoliation and bursts of He after a dose of 3 × 10¹⁷He⁺/cm². Rapid He evolution, with hole formation, was found above 0.7 T_m. No significant differences in either gas re-emission or surface deformation were found between the two fluxes employed.     

Surface Irradiation Damage of 316 Stainless Steel,Vanadium and Molybdenum by Helium Ion Bombardment

Molybdenum, V and 316 stainless steel were irradiated with 50~150 keV He ions at the temperatures between 413 and 1,298K for total doses ranging 1× 10²²~10ײ³ m^?2, and the characteristics of the surface damage were compared. Severe exfoliation was observed in all of these materials for the irradiation at 413±110 and 748±25K. The number of exfoliated skins was larger than that in literature, and increased nearly in proportion with the total dose. It increased in the order Mo<316SS<. When the dose was low, the amount of erosion increased rapidly with total dose, but tended to be saturated for higher doses than 3×10²² m^?2. It increased in the order Mo<V<316SS at 413±110K, while in the order 316SS<Mo<V at 748±25K. At higher temperatures than 923 K, blisters and porous surface were formed and the exfoliation of skins ceased. The amount of erosion increased with increasing incident ion energy in the energy range between 50 and 150 keV at 413±110K for a total dose of 1×10²² m^?2.     

Advanced Doped UO2 Pellets in LWR Applications

The nuclear industry strives to reduce the fuel cycle cost, enhance flexibility and improve the reliability of operation. This can be done by both increasing the fuel weight and optimizing rod internal properties that affect operational margins. Further, there is focus on reducing the consequences of fuel failures. To meet these demands Westinghouse has developed ADOPT (Advanced Doped Pellet Technology) UO₂ fuel containing additions of chromium and aluminium oxides. This paper presents results from the extensive investigation program which covered examinations of doped and reference standard pellets both in the manufactured and irradiated states.

The additives facilitate pellet densification during sintering and enlarge the pellet grain size. The final manufactured doped pellets reach about 0.5% higher density within a shorter sintering time and a five fold larger grain size compared with standard UO₂ fuel pellets. The physical properties of the pellets, including heat capacity, thermal expansion coefficient, melting temperature, thermal diffusivity, have been investigated and differences between the doped and standard UO₂ pellets are small.

The in-reactor performance of the ADOPT pellets has been investigated in pool-side and hotcell Post Irradiation Examinations (PIEs), as well as in the Studsvik R2 test reactor. The investigations have identified three areas of improved operational behaviour: Reduced fission gas release, improved PCI performance thanks to increased pellet plasticity and higher resistance against post-failure degradation. Fuel segments have been exposed to ramp tests and enhanced power steady-state operation in the Studsvik R2 reactor after base-irradiation to above 30 MWd/kgU in a commercial BWR. ADOPT reveals up to 50% lower fission gas release than standard UO₂ pellets. The fuel degradation behaviour has been studied in two oxidizing tests, a thermal-microbalance test and an erosion test under irradiation. The tests show that ADOPT pellets have a reduced rate of fuel washout, as compared to standard UO₂ pellets.

Fuel rods with ADOPT pellets have been irradiated in several light water reactors (LWRs) since 1999, including two full SVEA-96 Optima2 reloads in 2005.     

Some effects of pH on inorganic iodine volatility in containment

The behaviour of iodine in containment in the event of an accident involving fission product release would be strongly dependent on pH. High pH leads to a lower rate of radiolytic oxidation, and in alkaline conditions the thermally stable form is IO₃⁻. Much of the work on effects of pH on radiolytic oxidation reported in the literature may have been misinterpreted because of post-irradiation reaction and in this report some new experiments are described which were designed to overcome these problems involving sparged irradiated solutions of CsI spiked with ¹³¹I. The rate of radiolytic oxidation has been measured as a function of pH between pH 4.6 and pH 9 and iodide concentrations between 10⁻⁴ and 10⁻⁶ mol dm⁻³. Also discussed in the paper are factors which can affect the pH of the sump water and the effects of high pH in sprays. It is concluded that high pH is beneficial and it is important not only to achieve high pH but also to maintain it.     

Sodium Fires and Aerosols

Release of ¹³¹I associated with burning of contaminated Na, decay of airborne concentration of the released ¹³¹I and the size distribution of aerosols referred to radioiodine are investigated in experimental runs conducted on laboratory scale. These investigations are carried out in conjunction with similar investigations for the Na matrix. An experimental chamber (5.4 m³) is used for burning a small pool of Na (–50 g) containing spiked ¹³¹I (40–500 μCi). Values of the specific activity ratio, viz. the ratio of the concentration of ¹³¹I in the aerosol Na to that in pool Na lay in the range of 0.9×10^?2—5.7×10^?2 at pool temperatures of about 300°C. The concentration decay half-time and the aerosol size distribution characteristics referred to ¹³¹I remain similar to those applicable to the Na content of the aerosols. Surface concentrations of ¹³¹I in the residues examined differ from the bulk concentrations (calculated) in the pool and are lower by a factor of ?20. The concentrations of ¹³¹I in the aerosols released are further lower than the surface concentrations in the residues by a factor of ?2.